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The paper describes the main features of the Superconducting Magnets Power Supply to generate the toroidal and poloidal magnetic fields in JT-60SA tokamak, with special regard to coil current regulation mode and magnets protection.

A 170 GHz, 2 MW, steady-state gyrotron is being developed in collaboration between European research associations and industries to be used for the electron cyclotron resonance heating (ECRH) system of ITER, the gyrotron is presently in the prototype state. EFDA entrusted ENEA with the design of a new power supply for the gyrotron's body; in which, the traditional power vacuum tubes are replaced with solid-state components (IGBT); furthermore ENEA had the task of developing the new body power supply (BPS), following the initial conceptual design. The BPS is formed by 50 identical bidirectional (in current) modules, connected in series; the BPS's control system regulates the output voltage by changing the modulation index. OCEM Spa, under ENEA's supervision, has manufactured the BPS. Tests, done at the start of this year, showed the substantial agreement between the initial technical specifications and the values achieved. Particularly the ramp-up/down-emergency time (50–10 μs), the accuracy (±0.5%), the sine wave modulation (45–20 kV at 5 kHz) and the maximum phase shift ( In the article, the new stair-way modulation (SWM) scheme will be sketched and the test results will be described in detail.

The activities in Europe towards the development of the EC power sources for ITER are centered on the development of a 170GHz, 2MW, CW coaxial cavity gyrotron of collector potential depressed (CPD) type. A gyrotron with a higher unit power than the ITER reference (I MW) would yield a reduction of the installation costs, a more compact launcher design and, if required, an increase of the power delivered through one port. Tests proving the principle were successfully performed on a short-pulse experimental gyrotron delivering up to 2.2 MW in single mode. Following this success, a coordinated and fully consistent programme of development has been launched. The first industrial 2 MW prototype is now at an advanced stage of construction. The associated superconductive magnet producing 6.86 T on the cavity axis is also being procured. Dummy loads suitable for short and CW operation are also part of the development effort. Finally, a new EC test facility, with the features necessary for the testing of the gyrotron up to full power in CW, has been established and includes a fully solid-state power supply system. (c) 2007 Published by Elsevier B.V.

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load (P/R ∼ 22 MW/m; P/R2 ∼ 12 MW/m2) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ∼ 8.5 T; IP ∼ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ∼ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L–Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.

A conceptual design and feasibility study has been carried out for a fully solid state Body Power Supply providing the necessary performance required for the modulation of RF power from CW, high power gyrotrons. The selected solution, based on the Pulse Step Modulation concept, is expected to yield cost and operational advantages compared to systems based on vacuum tubes. A second phase for the manufacturing of a full size unit and its testing on a power gyrotron is now in progress.

In the frame of development of the ITER electron cyclotron wave (ECW) system, a two MW CW coaxial cavity gyrotron will be developed during the Sixth Framework Program (2003-2006). Such development relies on the availability of a test stand capable of providing the electrical energy and cooling capacity. This test stand will possibly be used, in a later stage, for the component test of the ITER ECW system. This paper will first present the main parameters of this new coaxial gyrotron. Then we describe the test stand itself, including the general requirements for testing and evaluating the behaviour of the RF source and then a description of the electrical system design. Compared to the ITER reference design, the test stand emphasises the requirement of flexibility, which is necessary during the development of the gyrotron. The additional electrical equipment is included in the overview of the electrical system. The cooling system will be an important part of the design study. Indeed, the design efficiency of such a depressed collector gyrotron is approximate to 50%, implying > 4 MW of continuous heat dissipation and evacuation by the cooling equipment. The specifications of the cooling system must also comply with ITER reference design values. (C) 2003 Published by Elsevier B.V.

The capability of the poloidal field coil system, as presently designed, to provide an effective vertical stabilization of the plasma in the IGNITOR machine has been investigated using the CREATE-L response model. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The growth rate of the vertical instability and the power required by the active stabilization system has been estimated with this model. The possible failure of the relevant electromagnetic diagnostics has been taken into account, evaluating the robustness of the plasma position reconstruction strategy. A realistic description of the power supply system has permitted to carry out the optimization of the proportional-integrative-derivative (PID) controller, both with a voltage and a current loop control scheme. An assessment of the requirements for the plasma cross section shape control has been carried out considering perturbations of the plasma global parameters independent of each other and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been rated relative to their capability to restore shape modifications due to different plasma disturbances. The most effective coil combination, that minimizes recovery time and voltage required, has been identified. (C) 2007 Elsevier B.V. All rights reserved.